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| Book/Report | FZJ-2016-04911 |
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1984
Kernforschungsanlage Jülich GmbH, Zentralbiliothek, Verlag
Jülich
Please use a persistent id in citations: http://hdl.handle.net/2128/12367
Report No.: Juel-1899
Abstract: The PUREX process has become the reprocessing procedure quite generally used for all fuel types containing natural, slightly or highly enriched uranium together with lower or higher contents of plutonium. The THOREX process on the other hand has been developed for reprocessing uranium and thorium from thorium-based fuel. When the fuel being irradiated contains appreciable amounts of $^{238}$U, the plutonium thus formed requires that a combination of the THOREX and PUREX processes must be applied. The THOREX process is technically less advanced and principally incriminated with the drawback that thorium nitrate exhibits a much lower distribution coefficient than uranium and plutonium. To drive thorium into the organic TBP phase, a strong salting agent is required. Aluminium nitrate, which formerly has been recommended is now replaced by nitric acid in order to reduce the amount of radioactive waste. However, high acid concentrations are counter-effective in achieving high fission product decontamination. Therefore, several flowsheet variants with acid and acid deficient feed solutions, respectively, have been investigated in the past. In order to achieve high decontamination factors, a dual cycle THOREX process was developed. This process uses an acid feed solution in the first cycle and an acid deficient one in the second cycle. According to recent investigations, a single cycle process with acid feed solution should provide the necessary decontamination factors. An immediate separation of thorium and uranium appears advisable in view of both fuel cycle strategy and process feasibility. Pulse columns should be preferentially used as extraction apparatus, at least for the xtraction step. Reprocessing of HTR-LEU fuel in existing PUREX plants is bothered with problems of criticality prevention due to the residual total fissile isotope content in the feed of greater than 2 %. Special precaution must therefore be taken in several processingunits. The addition of soluble poison to the feed proved to be an unfavourable measure. A more suitable method to overcome criticality difficulties may be the application of a flow sheet with lower TBP concentrations in the range of less than 10 %. Beyond it, there are no further problems to consider with LEU fuel reprocessing.
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